MC/DC: Monte Carlo Dynamic Code
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Updated
Sep 13, 2024 - Python
MC/DC: Monte Carlo Dynamic Code
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
Command line tool to convert MCNP mesh tallies to Visual ToolKit (VTK) formats. Supports all MCNPv6.2 legacy meshtal output formats, for both for rectangular and cylindrical meshes.
Stochastic Calculator Of Neutron transport Equation
OpenMC Monte Carlo Code
List of open source projects related to OpenMC
Energy-dependent neutron transport Monte Carlo implemented in Rust.
Meshing library for nuclear workflows
An open source utility to convert various publicly available macroscopic nuclear cross section formats
A modular toolkit of fast and reliable libraries for neutronics analysis. Several command line tools are built with this core collection of crates.
THOR is a radiation transport code for unstructured meshes.
Openmc-FEnicsx for muLtiphysics tutorIAl
A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools
An open source slide deck on fusion neutronics
A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
Simple code to simulate Neutron Scattering in Non-Radiative matter using Monte Carlo simulations
A pretty viewer for XSM files generated by DRAGON/DONJON or APOLLO neutronic codes
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